IR 05000302/1994009
| ML20029E084 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/10/1994 |
| From: | Butcher R, Cooper T, Landis K, Long A, York J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20029E082 | List: |
| References | |
| 50-302-94-09, 50-302-94-9, NUDOCS 9405160231 | |
| Download: ML20029E084 (21) | |
Text
'l UNIVED STATES
[cmstrop 'o NUCLEAR REGULATORY COMMisslON
.p" REGION 11
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j 101 MARIETTA STREET, N.W.
- t ATLANTA, GEORGI A 30323
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Report No.:
50-302/94-09 Licensee:
Florida Power Corporation 3201 34th Street, South St. Petersburg, FL 33733 Docket No.:
50-302 License No.:
DPR-72 facility Name:
Crystal River 3 Inspection Conducted: April 2 through May 6, 1994
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Inspector:
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c, tr y R. Butcher, Senior Resi~ dept Inspector Date Signed Inspector:
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T. Cooper, Resident Inspector Dat'e S'igned Inspector:
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J. York, Resident Inspecto'r, Surry Date S'igned Inspector:
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t 4. Long, Proj ct Engineer,9RII Date Signed Approved y:
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K. Lhnd,is, Section Chief Date/ Signed Divisidn of Reactor Projects SUMMARY Scope:
This routine inspection was conducted by the resident and regional inspectors in the areas of plant operations, radiological controls, security, fire protection, surveillance observations, maintenance observations, licensee event reports, and licensee action on-previous inspection items, licensee self assessment, and refueling activities. Numerous facility tours'were conducted and facility operations observed.
Backshift inspections were conducted on April 6, 7, 10, 20, 21, and 24.
9405160231 940509 PDR ADOCK 05000302 O
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Results:
Within the scope of this inspection, the inspectors determined that the licensee continued to demonstrate satisfactory performance to ensure safe plant operations.
One unresolved item ** was identified.
50-302/94-09-01, Unresolved Item. Insufficient Voltage to Operate Main Feedwater Isolation Valve FWV-28. (paragraph 4.a)
During this inspection period, the inspectors had comments in the following Systematic Assessment of Licensee Performance functional areas:
plant Operations:
The administrative controls during reduced reactor vessel level operations and TS refueling requirements were comprehensive and were considered a strength.
(paragraph 4.a and 10)
Maintenance / Surveillance:
The use of the ECAD 1000 System to assess and diagnose cable condition has improved troubleshooting and trending techniques.
This is considered a strength in the maintenance area.
(paragraph 6.a)
The licensee has shown a definite improvement in work controls and access controls for the switchyard area.
Corrective actions developed and being implemented in the switchyard to prevent recurrence of foul weather problems are considered to be a strength. (paragraph 6.b)
The failure to accomplish all known maintenance work items on the A EDG during the refueling outage and the scheduling of an on-line A EDG system outage in July 1994 is considered a weakness.
(paragraph 6.c)
Plant Support: (Radiation Controls, Emergency Preparedness, Security, Chemistry, Fire Protection, Fitness for Duty, and Housekeeping Controls)
The mock-up of the reactor vessel internal vent valve on the turbine deck provided for worker familiarization with the replacement task to help minimize exposure and is considered a strength.
(paragraph 4.b)
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- Unresolved items are a matter about which more information is required to determine whether they are acceptable or may involve violations or deviations.
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The following general comments were also noted:
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A visual examination of the reactor building sump indicated the sump screen assembly was in good condition. (paragraph 4.a)
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The large number of personnel contaminations that occurred beginning April 15, 1993, was inspected by Region II specialists and the results of that inspection will be documented in IR 50-302/94-10. (paragraph 4.b)
The inspectors reviewed the following outstanding items:
Item Number Stat n Description and R9ference VIO 50-302/92-27-02 Closed Failure to Enter an Action Statement With the EDG Inoperable as the Result of Non-TS Surveillance Testing.
(paragraph 9.a)
LER 92-02 Closed Shutdown Required by Technical Specification 3.0.5 Due to Inoperable Emergency Diesel Generator and Inoperable Vital Bus Transformer.
(paragraph 7.a)
LER 93-06 Closed Inadequate Modification Design Results in Degraded Seismic Capability and A Condition Outside the Design Basis.
(paragraph 7.b)
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REPORT DETAILS 1.
Persons Contacted Licensee Employees W. Bandhauer, Nuclear Shift Manager
- G. Boldt, Vice President Nuclear Production J. Campbell, Nuclear Shift Manager R. Davis, Manager, Nuclear Plant Maintenance
- G. Halnon, Manager, Nuclear Plant Operations B. Hickle, Director, Nuclear Plant Operations W. Marshall, Nuclear Shift Manager
- P. McKee, Director, Quality Programs
- R. McLaughlin, Nuclear Regulatory Specialist B. Moore, Manager, Nuclear Integrated Scheduling
- W. Neuman, Supervisor, Inservice Inspection
- W. Rossfeld, Manager, Site Nuclear Services W. Stephenson, Nuclu.r Shift Manager F. Sullivan, Nuclear Shift Manager
- R. Widell, Director, Nuclear Operations Site Support G. Wilson, Nuclear Shift Manager K. Wilson, Manager, Nuclear Licensing Other licensee employees contacted included office, operations, engineering, maintenance, chemistry / radiation, and corporate personnel.
NRC Resident Inspectors
- R. Butcher, Senior Resident Inspector
- T. Cooper, Resident Inspector Other NRC Personnel on Site S. Ebneter, Regional Administrator, RII i
J. Johnson, Acting Director, DRP, RII l
K. Clark, Public Affairs Officer, RII R. Trojanowski, Regional State liaison Officer, RII H. Berkow, Director, Project Directorate 11-2, NRR L. Raghavan, Project Manager, Project Directorate 11-2, NRR
- Attended exit interview Acronyms and initialisms used throughout this report are listed in the last paragrap.'
2.
Other NRC Inspections / Activities Performed During This Period REPORT NO.
Inspection Period ABEA INSPECTED l
50-302/94-02 April 13, 1994 SALP Report Presentation 50-302/94-10 April 18-21, 1994 Radiological Effluents and Chemistry 50-302/94-11 April 18-22, and Non-Destructive May 2-6, 1994 Examination and In-Service-Inspection On April 13,1994, at 1:00 p.m. a public meeting to discuss the CR-3 SALP report was held at the Licensee's Site Administration Building.
The SALP report results were presented by the NRC with the licensee
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given an opportunity to respond if desired. The SALP presentation i
covered the areas of Plant Operations, Maintenance, Engineering, and Plant Support for the time period of August 23 through February 19, 1994.
Following the SALP presentation on April 13, 1994, NRC management met with Local Officials. Discussions centered on the Local Officials stressing the importance of Emergency Preparedness and the continuation of Emergency Preparedness exercises.
3.
Plant Status At the beginning of this reporting period, Unit 3 was operating at approximately 78% power in preparation for the upcoming refueling outage and had been on line since September 20, 1994. On April 6, 1994, at 8:30 p.m., with the unit at 78% power, a power reduction was initiated in preparation for the refueling outage. The following evolutions occurred:
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April 7 at 1:48 a.m. - Main generator output breakers were opened.
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April 7 at 1:59 a.m. - Reactor power reduced to less than 5%.
Entered mode 2.
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April 7 at 3:20 a.m. - Entered mode 3.
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April 14 at 5:00 a.m. - Entered mode 6.
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April 22 at 3:26 p.m. - Initiated fuel off-loading.
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April 26 at 12:00 p.m. - Completed fuel off-loading.
4.
Plant Operations (71707 & 93702)
Throughout the inspection period, facility tours were conducted to observe operations and maintenance activities in progress. The tours included entries into the protected areas and the radiologically controlled areas of the plant. During these inspections, discussions were held with operators, health physics and instrument and controls technicians, mechanics, security personnel, engineers, supervisors, and plant management.
Some operations and maintenance activity observations
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were conducted during backshifts.
Licensee meetings were attended by the inspector to observe planning and management activities. The inspections confirmed FPC's compliance with 10 CFR, Technical Specifications, License Conditions, and Administrative Procedures.
a.
Operations
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Reduced Reactor Coolant System Inventory Operation In preparation for the CR-3 refueling outage (9R) scheduled to begin on April 7, 1994, at 2:00 a.m., when the generator output breakers were scheduled to be opened, the inspectors reviewed the licensee's administrative controls for operation of the RCS in reduced inventory and midloop conditions. AI-504, Guidelines for Mode 5 Outages and Reduced Reactor Coolant System (RCS) Inventory Operations, contains the following definitions:
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Reduced Reactor Vessel Inventory - A reduced RCS inventory condition when the level in the reactor vessel is <132 feet.
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Midloop Operations - A reduced RCS inventory condition when the level in the reactor vessel is <129 feet, 6 inches.
The reactor vessel level of 132 feet is three feet below the reactor vessel flange and a reactor vessel level of 129 feet, 6 inches is the top of the flow area of the hot legs at the junction with the reactor vessel.
Both of these definitions agree with GL 88-17 definitions.
Additionally, the licensee has the following definition:
Reduced RCS Inventory - Plant conditions during plant
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shutdown when the RCS is not filled and vented or the transfer canal is not filled to >156 feet.
Al-504 states that this procedure establishes guidelines that provide a level of safety above TS requirements for reduced RCS inventory conditions.
Enclosure 4 of Al-504, Justification for Departure From the Requirements of This Procedure, provides for departing from the requirements of AI-504 with adequate justification. The Director, Nuclear Plant Operations and the PRC must approve any deviations from the requirements of Al-504.
Procedures were verified to be active and ready for use for the following requirements when in reduced RV level or midloop operation.
(1)
Containment Closure Capability For Mitigation Of Radioactive Releases AI-504 requires in reduced RV level or midloop operation that the equipment hatch will normally be maintained in the l
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4 closed (unbolted) position except when required to be open for equipment movement to and from the RB. Additionally, in midloop operation, the equipment hatch will be temporarily installed with a minimum of 16 bolts per HP-114, Removal and Reinstallation of Equipment Hatch.
The status of other major containment penetrations (e.g.: equipment hatch air lock, personnel hatch air lock, and purge supply / exhaust valves) are carried on the Shutdown Daily Plant Status Report.
(2)
RCS Temperature - At least Two Independent, Continuous Indications That Are Representative Of Core Exit Conditions, Are Operable Al-504 states to record RCS level and temperature per Enclosure 2, RCS Level and Temperature Logsheet, of SP-301, Shutdown Daily Surveillance Log.
SP-301, Enclosure 2, requires RCS temperature be recorded every 15 minutes using 2 Incore temperatures when reactor vessel level is <129 feet, 6 inches. When reactor vessel level is less than 135 feet, record temperatures hourly using incore temperatures if available, otherwise use DHP suction temperature.
(3)
RCS Level Indication - At least Two Independent, Continuous Water Level Indications Operable (Calibrated)
Al-504 states to record RCS level and temperature per Enclosure 2, Level and Temperature Logsheet, of SP-301, Shutdown Daily Surveillance Log.
SP-301, Enclosure 2, requires reactor vessel level be recorded every 15 minutes from tygon tubing and from reactor vessel level instrumentation when reactor vessel is <132 feet, and hourly when reactor vessel level is <135 feet.
Reactor vessel level instrumentation that is to be used is RC-201-LII, or RC-201-LI2, or RC-202-LI, or computer point R329/R330.
(4)
RCS Perturbations should be Avoided Al-504 and OP-404, Decay Heat Removal System, include precautions to prevent or minimize RCS perturbations when-the reactor vessel is in reduced inventory or midloop operations.
(5)
RCS Inventory Addition - At least Two Additional Means Of Adding Inventory To The RCS Must Be Available, In Addition To The Pumps That Are Part Of The Normal RHR Systems -
Verify Operability
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AI-504 requires two available means of adding inventory to the RCS (in addition to the decay heat removal system) when the reactor vessel level is <132 feet. This includes at least one makeup pump and flow path through one injection val ve. Additional methods of adding inventory to the RCS, if available, are included as Enclosure 2 to AI-504.
(6)
Nozzle Dams / Loop Stop Valves - Reasonable Assurance Is Obtained That All Hot Legs Are Not Blocked Simultaneously Unless A Vent Path Is Established To Prevent Pressurization Of The Upper Plenum Of The Reactor Vessel This requirement does not apply to B&W NSSS.
CR-3 is a B&W facility with no loop stop valves.
(7)
Licensee Has Contingency Plans To Repower Vital Busses From Alternate Source If Primary Source Is Lost Enclosure 1 to AI-504 delineates the electrical power supply requirements for all modes of reduced inventory operation.
The primary power source (backfeed from the 500kV-yard) and one alternate 230kV power source (either the Offsite Power Transformer (MTTR-9) or the Backup ES Transformer (MTTR-6))
should be available at all times.
If the 500kV backfeed is established (through breaker 3207 or 3208) and one 230kV power source (MTTR-9 or MTTR-6) is available, only one operable EDG is considered acceptable.
In summary, the licensee's administrative controls contained in Al-504 placed system availability requirements above and beyond those outlined in GL 88-17 for operation in reduced inventory conditions and is considered a strength. The inspectors reviewed the licensee's responses to GL 88-17 and verified the licensee had reviewed their controls and administrative procedures for reduced reactor vessel inventory and midloop operation.
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On April 4, 1994, with the plant in mode 1 at 79% reactor power, while performing SP-321, Power Distribution Breaker Alignment and Power Availability Verification, the voltage on the A and B ES 4160V busses was 4350V. SP-321, Enclosure 2 and Enclosure 3 requires the ES 4160V busses be between 4050V and 4300V.
TS 3.8.9, Distribution Systems-0perating, SR 3.8.9.1 requires the licensee verify correct voltage to required AC vital bus electrical power distribution subsystems every 7 days.
SP-321 accomplishes this TS requirement.
TS 3.8.9 only addresses actions for the loss of one AC electrical power distribution subsystem, therefore the NSS entered TS 3.0.3 at 9:45 p.m. on April 4, 1994.
TS 3.0.3 requires when an LC0 is not met, that action be initiated within one hour to place the unit in mode 3 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.
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NSS contacted the System Dispatcher to have the grid voltage
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lowered. At 10:08 p.m. the B ES 4160V bus voltage was lowered to less than 4300V and TS 3.0.3 was exited. TS 3.8.9, Action A
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states that with one AC electrical power distribution subsystem inoperable, restore the AC electrical power distribution system to operable status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. At 10:10 p.m., both the A and B ES 4160V bus voltages were less than 4300V and TS 3.8.9 was exited.
Engineering is reviewing the SP-321 voltage criteria to determine if procedural changes should be made. The inspectors will follow-up on the licensee's corrective actions, if any, for this event.
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On April 11, 1994, the inspectors toured the RB and, as part of the tour, examined the condition of the RB sump and the sump screen assembly.
The CR-3 sump has a grating cover that has 1.5 inch openings that prevent large particles from entering the sump.
There is then a 3 foot weir wall to help segregate dense debris from the sump screen assembly. The sump screen is a 1/4 inch mesh wire. There are 2 fourteen inch diameter suction pipes from the RB sump to the DH system and the BS system. The RB sump configuration is described in FSAR Section 6.2, Reactor Building Spray System, and figure 6.6, Reactor Building Sump Screen Assembly.
The inspectors visual inspection indicated that the sump and sump screen assembly were in good condition with no holes or damaged areas visible.
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On April 12, 1994, at 11:40 a.m. with the reactor in mode 5, the licensee made a four hour notification to the NRC HQ Duty Officer per 50.72(b)(2)(i) regarding valve FWV-28. An improved modeling of the electrical properties affecting the motor voltage show that there would not be sufficient voltage at the terminals of the FWV-28 motor if accident conditions coincident with degraded grid conditions were present when FWV-28 was required to operate. A generic concern with respect to the operability of limitorque MOVs with the brake option installed was previously identified and documented in PR 93-0173. At low voltages, such as noted above, the motor would attempt to operate the valve open or closed, but may be prevented from doing so because of the inability of the brake to release.
FWV-28, along with other MOVs, were determined to be acceptable at that time with the information available.
TS 3.7.3, Main Feedwater Isolation Valves (MFIVs), (applicable only in modes 1, 2 and 3) requires two MFIVs in each MFW flow path be operable with at least one MFIV capable of isolating MFW within the required isolation time. The MFIVs are designated valves in the MFW system which function in conjunction with other equipment to isolate MFW to the OTSGs in accordance with assumptions used in the high energy line break accident analyses.
At CR-3, the MFIVs for each OTSG consist of the MFW pump suction valve (FWV-14 or FWV-15), the main (FWV-29 or FWV-30)/startup (FWV-33 or FWV-36)/ low load (FWV-31 or FWV-32) block valves (in parallel), and the MFW pump discharge cross connect valve (FWV-28)
between OTSG A and B.
All the OTSG A valves receive a signal to close on low OTSG pressure from EFIC OTSG A MFW isolation j
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automatic actuaticn logic channels A and B.
OTSG B valves similarly receive signals from EFIC OTSG B MFW isolation automatic actuation logic channels A and B.
The crossover valve receives closure signals from both channels of EFIC's OTSG A and 0TSG B MFW isolation logics.
Since the reactor was in mode 5, no TS action was applicable. The licensee initiated PR 94-0087 to document this problem. The inspectors will follow up on the licensee's corrective actions prior to unit restart from the refueling outage. This issue will be tracked as Unresolved ! tan 50-302/94-09-01, Insufficient Voltage to Operate Mair. Feedwater Isolation Valve FWV-28.
b.
Radiological Protection Program Radiation protection control activities were observed to verify that these activities were in conformance with the facility policies and procedures, and in compliance with regulatory requirements.
These observations included:
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Entry to and exit from contaminated areas, including step-off pad conditions and disposal of contaminated clothing;
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Area postings and controls;
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Work activity within radiation, high radiation, and contaminated areas;
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RCA exiting practices; and
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Proper wearing of personnel monitoring equipment, protective clothing, and respiratory equipment.
The implementation of radiological controls observed during this inspection period were proper and conservative.
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On April 15, 1994, at approximately 7:00 p.m., personnel exiting the RB were alarming the RB exit PCMs and the RCA exit PCMs, indicating skin contamination.
Further investigation indicated the contamination was due to the presence of I-132.
PR 94-0098 was initiated to document this problem. A specialist inspector from the NRC Region II office arrived on April 18, 1994, to follow-up this event.
See NRC IR 50-302/94-10 for details.
The resident inspectors will follow the licensee's resolution of PR 94-0098.
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During video inspection of the reactor vessel internal vent valves, the jack screw locking cup on RCV-170 was identified as missing. There are two jack screws per valve which keep the valve located in the core support shield. The locking cup prevents the jack screws from rotating, therefore keeping the valve properly located. The licensee replaced the damaged valve with a new valve. Work was performed using a long handled tool designed to remove the valve from the core support shield. The new valve wac then installed and tested. The damaged valve was transported from the RB to the spent fuel floor and placed in the Decon Pit.
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the outage, the remaining portion of the jack screw locking cup will be removed from the valve and a failure analysis performed.
To keep exposures ALARA, the licensee obtained an actual vent valve and installed it in a simulated core support shield for training to familiarize the workers with the actual job.
The mock-up was placed on the turbine deck with appropriate scaffolding so the long handled tool could be used for familiarization.
The mock-up for worker familiarization with potentially high exposure jobs is a good practice and is considered a strength.
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c.
Security Control In the course of the monthly activities, the inspector included a review of the licensee's physical security program.
The performance of various shifts of the security force was observed in the conduct of daily activities to include:
protected and vital areas access controls; searching of personnel, packages, and vehicles; badge issuance and retrieval; escorting of visitors; patrols; and compensatory posts.
In addition, the inspector
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observed the operational status of protected area lighting, protected and vital areas barrier integrity, and the security organization interface with operations and maintenance. No performance discrepancies were identified by the inspectors.
d.
Fire Protection Fire protection activitie:, staffing, and equipment were observed
'to verify that fire brigade staffing was appropriate and that fire alarms, extinguishing equipment, actuating controls,. fire fighting equipment, emergency equipment, and fire barriers were operable.
Violations or deviations were not identified.
One unresolved item was identified.
5.
Surveillance Observations (61726)
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The inspectors observed TS required surveillance testing and verified that the test procedures conformed to the requirements of the TSs; testing was performed in accordance with adequate procedures; test instrumentation was calibrated; limiting conditions for operation were met; test results met acceptance _ criteria requirements and were reviewed by personnel other than the individual directing the test; deficiencies -
were identified,.as appropriate, and were properly reviewed and resolved by management personnel; and system restoration was adequate.
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completed tests, the inspectors verified testing frequencies were met ~
and tests were performed by qualified individuals.
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The inspectors witnessed / reviewed portions of the following test activities:
- SP-179C, Containment Leakage Test - Type C Revision 11 (Penetration 116);
- SP-406, Refueling Operations Daily Data Requirements (Revision 19);
- SP-435, Valve Testing During Cold Shutdown (FWV-29 and FWV-30 Stroke Tests); and
- SP-650, ASME Code Safety Valves Test.
The inspectors determined that the above testing activities were performed in a satisfactory manner and met the requirements of the TSs.
On April 26, 1994, a CAL was sent to the licensee regarding the OTSG tube inspection for refuel 9 outage. This CAL recognized the unique criteria required to disposition tube indications that are not addressed by the TSs. A specialist inspector from Region II was on site to inspect this activity and this will be doc mented in IR 50-302/94-11.
Violations or deviations were not identified.
6.
Maintenance Observations (62703)
Station maintenance activities of safety-related systems and components were observed and reviend to ascert.ain they were conducted in accordance with approved procedures, regulatory guides, industry codes and standards, and in conformance with the TSs.
The following items were considered during this review, as appropriate:
LCOs were met while components or systems were removed from service; approvals were obtained prior to initiating work; activities were accomplished using approved procedures and were inspected as applicable; procedures used were adequate to control the activity; troubleshooting i
activities were controlled and repair records accurately reflected the inaintenance performed; functional testing and/or calibrations were performed prior to returning components or systems to service, QC records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were properly implemented; QC hold points were established and observed where required; fire prevention controls were implemented; outside contractor force activities were controlled in accordance with the approved QA program; and housekeeping was actively pursue,
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The inspectors witnessed / reviewed portions of the following maintenance activities in progress:
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WR 0318247, Perform Testing of Various Cables in the RPS Cabinets; and
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WR 314694, LHC Test Group 1 and Test Group 2 Push Butte s Did Not Stroke Their Associated Valves.
The following items were considered noteworthy.
a.
The inspectors witnessed the use of the ECAD System 1000, used to measure cable integrity in electrical systems. At the time of the use, the licensee was conducting tests on system cabling to the RPS control panels.
The ECAD 1000 system is an integrated, computer based system which measures resistance, inductance, and capacitance of the cables being tested.
Both AC and DC cable resistance and insulation resistance is measured. Time-domain reflectometry analysis and polarization index calculations are performed by the equipment.
Polarization index is defined as the ratio of the insulation resistance value measured at ten minutes to the insulation resistance value measured at one minute.
This value is also used to determine the condition of the insulation.
The field equipment has the capacity to compare the current measurements with a reference value, usually the results of an earlier test.
Data from this test can be used by a larger program, ECAD 2000, to perform more in depth analysis. The use of this system results in a faster, more efficient method of verifying safety-related cable integrity.
The inspectors witnessed troubleshooting of the EHC system using the ECAD 1000 system. Originally, it was thought that the' control solenoid in the circuitry was responsible for the problems with testing.
The TDR trace remained almost the same with the coil connected and disconnected. This would imply that the problem was not in the coil. The TDR trace shows a fairly constant impedance for about 3.75 feet after the incident pulse, then there was a constantly decreasing impedance distributed along the rest of the line.
The data pointed to a grounded conductor in the first 3.75 feet of the run.
This cable is run into the cabinet with several extra feet coiled in the cabinet. When the cable bundle with the bad conductor was freed from the uni-strut, the ground disappeared.
The conductor was then enclosed in Ray-Chem shim stock for the entire run in the
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cabinet. The conductor and coil then meggered >999 Meg ohms.
The ECAD system prevented the licensee from having to replace the i
entire conductor, which would involve many hours of preparation
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and cable pulling. The use of the ECAD 1000 System to assess and diagnose cable condition has improved troubleshooting and trending techniques. This is considered a strength in the maintenance area.
b.
Following the storm of March 1993 and the problems that developed with the switchyard during that storm, the licensee developed a list of corrective actions in an attempt to prevent a recurrence of the situation. Among these corrective actions were:
replace all 230kV cap and pin insulators with long leakage type post insulators, use non-stick coating on all 500kV and 230kV switchyard and plant electrical insulators, and replace conventional gap type lightning / surge arresters with modern metal oxide design at critical buses.
The inspectors toured the 230kV switchyard and observed the silicone based coatings being applied to the insulators to protect them from the salt buildups that were encountered during March 1993. The inspectors noted that switchyard access was controlled, with the Shift Supervisor having to grant permission for entry and the licensee workers actually controlling the keys to the gate lock.
Work activities were controlled by the licensee, but were performed by the coating vendor.
The licensee had two relay technicians and one relay supervisor present during the work, to oversee the contractor who actually applied the coatings.
While this work was being performed, the area of the switchyard being worked on was deenergized and was unavailable to supply emergency loads. The BEST transformer, which is located adjacent to the turbine building inside the protected area, was being used to supply the vital loads. Since work was being performed in its general vicinity, although not on the BEST transformer, a wooden barrier was erected around the transformer to reduce risk to the BEST transformer.
The inspectors noted that the work oeing performed in the switchyard was well planned and controlled, minimizing threats to the required power supplies. The licensee has shown a definite improvement in work controls and access controls for the switchyard area. Corrective actions developed and being implemented in the switchyard to prevent recurrence of foul weather problems are considered to be a strength, c.
During the present refueling outage, the A EDG was disassembled for inspection and maintenance work activities. This EDG outage lasted from approximately April 2 to April 29, 1994. An on-line system-outage for the A EDG has been scheduled for July 12 through July 14,1994, to perform miscellaneous corrective and preventive maintenance work items. Most, if not all, of these work scope items could have been accomplished during the refueling outage l
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when the A EDG was already disassembled. This failure to accomplish all known outstanding maintenance work items on the A EDG during the refueling outage is considered poor practice and a weakness.
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for those maintenance activities observed, the inspectors determined that the activities were conducted in a satisfactory manner and that the work was properly performed in accordance with approved maintenance work orders.
Violations or deviations were not identified.
7.
Onsite Follow-up and In-Office Review of Written Reports of Non-routine Events and 10 CFR Part 21 Reviews (90712/90713/92700)
The Licensee Event Reports and/or 10 CFR Part 21 Reports discussed below were reviewed.
The inspectors verified that reporting requirements had been met, root cause analysis was performed, corrective actions appeared appropriate, and generic applicability had been considered.
Additionally, the inspectors verified the licensee had reviewed each event, corrective actions were implemented, responsibility for corrective actions not fully completed was clearly assigned, safety questions had been evaluated and resolved, and violations of regulations or TS conditions had been identified. When applicable, the criteria of 10 CFR Part 2, Appendix C, were applied.
a.
(Closed) LER 92-02:
Shutdown Required by Technical Specification 3.0.5 Due to Inoperable Emergency Diesel Generator and Inoperable Vital Bus Transformer.
Requirements to check and routinely replace the seals on various support equipment to the EDGs were incorporated in the licensee procedure, SP-605, Emergency Diesel Generator Engine Inspection / Maintenance, Revision 30, effective February 25, 1993.
The inspectors verified that the revision was completed and was technically accurate.
In addition, the inspectors verified that the inspection of similar seals on other support equipment for the EDGs was completed and that the seals were replaced and were in acceptable condition.
Seals that had been replaced within the recent past and were in good condition were left in the components and will be included in the performance of SP-605. This LER is closed, b.
(Closed) LER 93-06:
Inadequate Modification Design Results in
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Degraded Seismic Capability and A Condition Outside the Design Basis, i
In 1990, a plant modification replaced flow indicators directly above the separation barrier for control switches fer MUV-23, MUV-24, MUV-25 and MUV-26. The flow indicators were of a new design and did not have provisions to secure the separation barriers as
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did the previous design. MAR 91-08-26-01 modified the separation i
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barriers installation to attach the barriers directly to the control board. The inspectors observed the separation barrier mounting modification being installed.
No further action is necessary. This LER is closed.
c.
On April 14, 1994, the resident inspectors received a copy of a letter dated April 4, 1994, from B&W Nuclear Technologies to Duke Power Company regarding PSC 1-94 on the Reactor Cavity Seal Plate.
The concern involved the effect of a possible impact of the seal plate on the CRDM Service Support Structure during a postulated core flood line LOCA.
It is postulated that during this event, the seal plate, which is stored during plant operation without hold down attachments immediately above the reactor cavity, could be lifted and rotated by the asymmetric cavity pressure and impact the CRDM Service Support Structure with a lateral force.
In effect, the seal plate would become a missile. This occurrence
could result in deflection of the Service Support Structure and
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interference with insertion of the control rods. This potential problem was previously analyzed in 1989 by B&W Nuclear Technologies for TMI-1. At that time, this issue was not considered a valid problem.
However, recent reinvestigation has revealed a deficiency in the original B&W Nuclear Technologies
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l analysis and it is now considered that the impact of the seal plate against the CRDM Service Support Structure could result in impeding the required insertion of the control rods and could also result in loss of the CRDM pressure boundary.
The inspectors examined the possible impact of this revised B&W Nuclear Technologies finding on CR-3.
By MAR 92-08-01-01, l
Permanent Canal Seal Plate, CR-3 installed a new fuel transfer i
canal seal plate of a design to be welded to the gasket seating i
surfaces at the outside diameter of the reactor vessel seal ledge l
and to the existing liner plate of the fuel transfer canal.
Flow openings were provided for cavity ventilation during normal operations.
Bolt on covers with 0-ring seals are installed for operation where the fuel transfer canal is flooded. The safety evaluation of MAR 92-08-01-01 re-examined the PSC 1-94 concern and t
i determined that the welded seal plate would not compromise the
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structural integrity of the service structure and the CRDMs.
MAR 92-08-01-01 was installed during the 1993 mid-cycle outage and was completed on April 20, 1993.
Based on this review, the inspectors determined that the PSC 1-94 regarding the reactor cavity seal plate was not applicable to CR-3.
Violations or deviations were not identified.
8.
Follow-up on Items of Non-compliance (92702)
A review was conducted of the following non-compliances to assure that corrective actions were adequately implemented and resulted in conformance with regulatory requirements.
Verification of corrective
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action was achieved through record reviews, observation, and discussions with licensee personnel.
Licensee corraspondence was evaluated to ensure the responses were timely and corrective actions were implemented within the time periods specified in the reply.
a.
(Closed) Violation 50-302/92-27-02:
Failure to Enter an Action Statement With the EDG Inoperable as the Result of Non-TS Surveillance Testing.
As discussed in IR 50-302/92-27 and the licensee's response, following the implementation of the improved TS, the licensee would change their practice and would enter TS Action Statements during the performance of TS required surveillance activities.
The improved TS were implemented on March 12, 1994. The licensee's commitment was demonstrated on March 15, 1994, when performing SP-340F, MVP-Ic and Valve Surveillance on MVP-1C and valves MVV-25 and MVV-26. This event was documented in paragraph 3.a of IR 50-302/94-07. This item is closed.
Violations or deviations were not identified.
9.
Self Assessment (40500)
The licensee routinely performs Quality Program audits of plant activities as required under its QA program or as requested by management To assess the effectiveness of these licensee audits, the inspectors examined the status, scope, findings and recommendations of the following audit report:
REPORT NO.
TITLE N0. OF NO. OF FINDINGS RECOMMENDATIONS 94-01-FFD Fitness For Duty
8 The following item was considered noteworthy.
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The licensee's audit on Fitness For Duty was conducted from January 17 thru January 31, 1994. A recent NRC inspection was conducted in the area of the Fitness For Duty Program and the results are documented in IR 50-302/94-06.
The dates for the NRC inspection were February 28 through March 3, 1994. The NRC IR stated a notable strength in the Fitness For Duty Program was the licensees audits of its collection facilities, contract testing facility, and contractors' programs.
No additional NRC follow-up will be taken on the finding referenced above because it was identified by the licensee's audit program and corrective actions have either been completed or are currently underway.
A PR was initiated on the finding and plant management is aware of the finding.
Violations or deviations were not identifie.
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10.
Refueling Activities (60710)
On April 22, 1994, at 3:26 p.m. the licensee began defueling operations.
The fuel transfer system at CR-3 consists of two transfer tubes which are serviced by two carriages each containing one fuel assembly basket.
At both ends of each carriage's travel, in the SFP and RB canal, a basket lift frame (upender) is provided to raise the fuel assembly for handling or to lower.the fuel assembly for transfer.
The two upenders are identified as the X (North) or Y (South) upender. The initial fuel off-load was initiated with the Y upender while the X upender was undergoing repairs. The inspectors used the following documents in verifying refueling activities met requirements.
- SP-406, Refueling Operations Daily Data Requirements, Revision 19; and
- FP-203, Defueling and Refueling Operations, Revision 32.
Preparations for refueling were previously inspected and documented in IR 50-302/94-05 and IR 50-302/94-07.
TS 3.9, Refueling Operations, was verified to have been met by the conduct of SP-406 (or other documents) as fallows:
TS Reauirement Acolicable Document 3.9.1, Baron Concentration SP-406, Paragraph 3.7.3 (SR 3.9.1.1):
Boron Concentrations of and Enclosure 1, the RCS and the refueling canal shall Section C be maintained within the limit specified in the COLR.
3.9.2, Nuclear Instrumentation SP-406, Enclosure 1, (SR 3.9.2.1): Two source range Section A neutron flux monitors shall be operable.
3.9.3, Containment Penetrations Shutdown Daily Plant (SR 3.9.3.1):
Containment Status Report penetrations shall be maintained in the required status.
3.9.4, Decay Heat Removal and Coolant SP-406, Enclosure 1, Circulation - High Water Level Section A (SR 3.9.4.1):
One DHR loop shall be in operation at a flow rate of 22700 gpm.
3.9.5, Decay Heat Removal and Coolant SP-406, Enclosure 1, Circulation - Low Water Level Section A (SR 3.9.5.1 & 3.9.5.2): Two DHR loops shall be operable and at least one DHR loop shall be in operation,
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3.9.6, Refueling Canal Water Level SP-406, Enclosure 1, (SR3.9.6.1):
Refueling canal water Section C level shall be maintained 2156 ft.
p1 ant datum.
Procedure SP-406 was accomplished daily by operations and the Shutdown Daily Plant Status Report was issued daily.
The inspectors routinely reviewed both reports for completeness and accuracy.
SP-406 was considered comprehensive and is considered a strength. No discrepancies were identified.
The inspectors witnessed fuel off-loading to the SFP and followed the logging of fuel movement and fuel tag board updates in the Control Center.
The fuel movement was conducted in a professional and well controlled manner with good communications between the Control Center, the refueling floor (inside the RB), and the spent fuel pool area. The inspectors verified the controls of refueling procedure FP-203 were adhered to.
Violations or deviations were not identified.
11.
Exit Interview The inspection scope and findings were summarized-on May 6,1994, with those persons indicated in paragraph 1.
The inspectors described the areas inspected and discussed in detail the inspection results listed below.
Proprietary information is not contained in this report.
Dissenting comments were not received from the licensee.
Item Number Status Descriotion and Reference VIO 50-302/92-27-02 Closed Failure to Enter an Action Statement With the EDG Inoperable as the Result of Non-TS Surveillance Testing.
(paragraph 9.a)
URI 50-302/94-09-01 Open Insufficient Voltage to Operate MFW Isolation Valve FWV-28.
(paragraph 4.a)
LER 92-02 Closed Shutdown Required by Technical Specification 3.0.5 Due to Inoperable Emergency Diesel Generator and Inoperable Vital Bus Transformer.
(paragraph 7.a)
LER 93-06 Closed Inadequate Modification Design Results in Degraded Seismic Capability and A Condition
>
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Outside the Design Basis.
(paragraph 7.b)
12.
- Alternating Current AI
- Administrative Instruction a.m.
- ante meridiem ASME - American Society of Mechanical Engineers BS
- Building Spray B&W
- Babcock & Wilcox CAL
- Confirmatory Action Letter CFR
- Code of Federal Regulations COLR - Core Operating Limit Report CR-3 - Crystal River Unit 3 CRDM - Control Rod Drive Mechanism
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- Direct Current DH
- Decay Heat Closed Cycle Cooling DHR
- Decay Heat Removal DHP
- Decay Heat Pump DRP
- Division of Reactor Projects EDG
- Emergency Diesel Generators EFIC - Emergency Feedwater Initiation and Control EHC
- Electro-Hydraulic Control ES
- Engineered Safeguards FFD
- Refueling Procedure FPC
- Florida Power Corporation FSAR - Final Safety Analysis Report FWV
- Feedwater Valve GL
- Generic letter
- gpm
- gallons per minute HQ
- Headquarters
- Iodine IR
- Inspection Report kV
- kilovolt LCO
- Limiting Condition for Operation LER
- Licensee Event Report MAR
- Modification Approval Record MFIV - Main Feedwater Isolation Valve MFW
- Main Feedwater M0V
- Motor Operated Valve MP
- Maintenance Procedure MVP
- Make-up Pump MUV
- Make-up Valve NRC
- Nuclear Regulatory Commission NRR
- Office of Nuclear Reactor Regulation NSS
- Nuclear Shift Supervisor NSSS - Nuclear Steam System Supplier OP
- Operating Procedure OTSG - Once Through Steam Generator PCM
- Portable Contamination Monitor
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18 p.m.
- post meridiem PR
- Problem Report PSC
- Preliminary Safety Concern QC
- Quality Control QA
- Quality Assurance RB
- Reactor Building RCS
- Reactor Coolant Valve RHR
- Reactor Protection System RV
- Reactor Vessel SALP - Systematic Assessment of Licensee Performance SFP
- Spent Fuel Pool SP
- Surveillance Procedure SR
- Surveillance Requirement TDR
- Time Domain Reflectometry THI
- Three Mile Island TS
- Technical Specification URI
- Unresolved Item V
- Volt VIO
- Violation WR
- Work Request J
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